D. R. Prabhu

Demonstration of T2EHDGA Based Process for Actinide Partitioning Part II: Counter-Current Extraction Studies

Abstract Counter-current mixer-settler studies for actinide partitioning were carried using N,N,N′,N′-tetra-2-ethylhexyl diglycolamide (T2EHDGA) as the extractant. The feed solution was Simulated High Level Waste of Pressurized Heavy Water Reactor (PHWR-SHLW) origin spiked with 241Am, 244Cm, 152Eu, 137Cs, 85,89Sr, 59Fe, 106Ru, 109Pd, 95Zr, and 99Mo tracers. The organic stream was 0.1 M T2EHDGA + 5% isodecanol in n-dodecane. Extraction, scrubbing, and stripping experiments were performed by maintaining an organic to aqueous phase ratio of 1. More than 99.9% of the trivalent actinides and lanthanides were extracted in four stages, and the decontamination factors (D.F.) values were >103 obtained for most fission products. The co-extraction of Zr and Pd was prevented by the addition of oxalic acid and N-(2-hydroxyethyl)-ethylenediamine-triacetic acid (HEDTA) into the feed solution. However, ∼20% Ru and 10% Mo was extracted into the organic phase, which was successfully scrubbed using a mixture of 0.2 M oxalic acid and 0.1 M HEDTA in 5 M HNO3. Finally, the extracted actinides and lanthanides were quantitatively stripped with 0.2 M HNO3. Raffinate of the extraction cycle was found to be free from any alpha activity.

Separation Studies of Uranium and Thorium Using Di-2-Ethylhexyl Isobutyramide (D2EHIBA)

The extraction behavior of di-2-ethylhexyl isobutyramide (D2EHIBA) in dodecane medium for U(VI), Th(IV), and fission products such as Zr, Ce, Eu, and Cs, and the structural material Fe, has been investigated over a wide range of nitric acid concentrations. It has been observed that whereas D U varies from < 10−3 (pH 2.0) to 4.4 (6 M HNO3) with 1 M ligand, the corresponding D Th values are 1.5 × 10−3 and 4 × 10−2. In the presence of 250 g/L of Th, D U values are 8.6 (pH 2.0) and 2.2 (6 M HNO3). D2EHIBA has been found to be a promising extractant of trace concentrations of U in the presence of macro amounts of Th. The extraction of fission products and Fe is found to be negligible. D2EHIBA is found to extract nitric acid predominantly as a 1:1 species (K H = 0.156 ± 0.048). U(VI) is extracted as a disolvate UO2(NO3)·2D2EHIBA (K ex = 0.87 ± 0.08).

EVALUATION OF DI(2-ETHYLHEXYL)ISOBUTYRAMIDE (D2EHIBA) AS A PROCESS EXTRACTANT FOR THE RECOVERY OF 233U FROM IRRADIATED Th

ABSTRACT Extraction behaviour of U(VI) and Th(IV) has been studied employing 0.5 M D2EHIBA (di(2-ethylhexyl)isobutyramide) in n-dodecane as a function of nitric acid concentration. An improved separation of U from Th could be achieved in the presence of fluoride ion. Use of 1 M Al(NO3)3 in 2 M HNO3 has been suggested as the scrubbing solution. Various physico-chemical properties like density, viscosity, interfacial tension (IFT) and limiting organic concentration (LOC) of Th/U have been measured to evaluate the use of D2EHIBA as a process solvent. Degradation behaviour (acid hydrolysis as well as gamma radiolysis) of D2EHIBA under process conditions has been investigated using potentiometry as well as distribution studies.

Di(2-ethyl hexyl) butyramide and di(2-ethyl hexyl)isobutyramide as extractants for uranium(VI) and plutonium(IV)

The extraction of uranium(VI) and plutonium(IV) from nitric acid into n-dodecane was studied using two isomeric branched alkyl amides, di(2-ethyl hexyl) butyramide (DEHBA) and di(2-ethyl hexyl) isobutyramide (DEHIBA). The extraction ratios of Pu(IV) at relatively high acidities were higher than the corresponding values for U(VI) in the case of DEHBA. However, with DEHIBA the values for Pu(IV) were negligibly small. Pu(IV) was found to be extracted as trisolvate by DEHBA and as disolvate by DEHIBA. U(VI) was extracted by both the amides. From the study of the extraction reactions at different temperatures, it was shown that all the reactions in the present investigation were enthalpy favoured and entropy disfavoured. Separation of Pu(IV) from bulk of U(VI) was feasible. However, the purity of the separated plutonium was not satisfactory in batch extraction studies.

Distribution studies on Th(IV), U(VI) and Pu(IV) using tri-n-butylphosphate and N,N-dialkyl amides

Summary Distribution studies on U, Pu and Th have been carried out employing tri-n-butyl phosphate (TBP) as extractant to optimise the conditions for the reprocessing of spent fuel of proposed Th based Advanced Heavy Water Reactor (AHWR). Due to their high basicity and organophilicity straight chain N,N-dialkyl amides have been evaluated for their third phase formation behaviour (usually quantified by limiting organic concentration) during the extraction of thorium vis a vis TBP. Dihexyl derivatives of octanamide (DHOA) and decanamide (DHDA) were found promising in view of higher limiting organic concentration of thorium viz. 33 g and 50 g/L, respectively as compared to 26 g/L for TBP. Extraction profiles of U, Pu and Th have been obtained using these amides and compared with those of TBP. Attempts have also been made to optimise the stripping conditions using aceto-hydroxamic acid (AHA) and acetaldoxime as reductants for the partitioning of Pu from the loaded organic phase.

TETRA-BUTYL-MALONAMIDE AND TETRA-ISOBUTYL MALONAMIDE AS EXTRACTANTS FOR URANIUM(VI) AND PLUTONIUM(IV)

ABSTRACT Two new symmetrical diamides, namely straight-chain alkyl substituted neutral tetra-butyl-malonamide(TBMA) and sterlcally hindered branched-chain alkyl substituted tetra-isobutyl malonamide(TIBMA) were synthesised, characterised and used for the extraction of U(VI) and Pu(IV) from nitric acid media into n-dodecane. Both the cations were found to be extracted as their disolvates. Interestingly TBMA extracted more efficiently than TIBMA but afforded poor selectivity for Pu/U separation. The thermodynamic parameters involved in the extraction, determined by the temperature variation method, indicated the reactions in all cases to be enthalpy favoured. Entropy was found to be counteracting the extraction of U(VI) and favouring the extraction of Pu(IV). The recovery of diamides from the loaded actinides could be easily accomplished by using dilute oxalic acid or dilute U(IV) as the strippant for Pu(IV) and using dilute Na2C03 as that for U(VI).

Extraction of Uranium(VI) and Plutonium(IV) with Unsymmetrical Monoamides

Solvent extraction of uranium(VI) and plutonium(IV) has been investigated by three unsymmetrical monoamides, viz., N-methyl, N-butyl derivatives of hexanamide (MBHA), octanamide (MBOA) and decanamide (MBDA) in n-dodecane at 25, 30, 40 and 50 °C (±0.1). The extraction of the metal ions follows the order of the basicity of the amides (except for plutonium (IV)-MBDA system) as indicated by the equilibrium constant for the uptake of nitric acid by the amide (KH). The order of the basicity was found to increase with the length of the carbon chain, viz., MBHA < MBOA < MBDA. Uranium(VI) and plutonium(IV) were found to be extracted as disolvates and trisolvates respectively, as was observed in the case of symmetrical monoamides. The thermodynamic parameters determined by the temperature coefficient method indicate the extraction reactions to be predominantly enthalpy stabilized.

Extraction of uranium(VI) and plutonium(IV) with some high molecular weight aliphatic monoamides from nitric acid medium

The extraction behavior of U(VI) and Pu(IV) with dioctyloctanamide (DOOA), dioctylethylhexanamide (DOEHA) and diisobutylethylhexanamide (DIBEHA) was investigated from nitric acid medium. With DOOA, U(VI) extraction is higher than that for Pu(IV) upto 5M HNO3 and the trend is reversed at higher acid concentrations. Extraction yield of U(VI) is higher than that for Pu(IV) in the case of DOEHA and DIBEHA. DIBEHA extraction of Pu(IV) is found to be very small. The lower value of the distribution ratio for Pu(IV) with branched amides was attributed to steric reasons. The possibility of using these amides for separation of U(VI) and Pu(IV) without valency adjustment was explored. Both U(VI) and Pu(IV) are extracted as their disolvates by DOOA and DOEHA.

DISTRIBUTION BEHAVIOUR OF U(VI), Th(IV) AND Pa(V) FROM NITRIC ACID MEDIUM USING LINEAR AND BRANCHED CHAIN EXTRACTANTS

ABSTRACT The extraction behaviour of 1M solutions of tri-2-ethylhexyl phosphate (TEHP), di-2-ethyl hexyl isobutyramide (D2EHIBA), tri-n-butyl phosphate (TBP) and di-n-hexyl hexanamide (DHHA) in n-dodecane towards U(VI), Th(IV) and Pa(V) in the presence of 220 g/L of Th from nitric acid medium has been studied. The limiting organic concentrations (LOC) of thorium (g/L) for 1 M TBP and 1 M DHHA are evaluated as 31, 20 ( at 1 M HNO3) and 25,13 (at 4 M HNO3) respectively. The distribution ratio (D) values of U(VI), Th(IV) and Pa(V) in the presence of thorium (220 g/L) at. 1 M HNO3 suggest that branching in the alky group of amides suppresses the extraction considerably. In view of the selective extraction of U over Th by 5 % TBP in THOREX process at 4 M HNO3, distribution behaviour is also studied employing a lower concenfration (0·18 M) of extractant for comparison purpose, Separation factor (S. F.) values for U(VI) over Th(IV) under different experimental conditions consistently varied in the order: D2EHIBA > DHHA > TEHP > TBP. The quantitative extraction of 233U from a synthetic mixture containing 233U (10−5 M). 233Pa (10−11 M) and thorium (220 g/L) at 1 M HNO3 using 1 M solution of D2EHIBA in n-dodecane is achieved in three stages, Stripping and reusability studies of D2EHIBA have also been carried out.

Recent R&D Studies Related to Coprocessing of Spent Nuclear Fuel Using N,N-Dihexyloctanamide

Abstract The PUREX process has undergone several modifications to address the issues of high burn up, fewer solvent extraction cycles, and reduced waste arisings. Advanced fuel cycle scenarios have led to a renewed international interest in the development of separation schemes for co-recovering U/Pu from spent fuels. Completely incinerable N,N-dihexyloctanamide (DHOA) has been identified as a promising candidate for the reprocessing of spent fuels. Batch extraction studies were carried out to evaluate DHOA and TBP for the coprocessing (co-extraction and co-stripping) of U and Pu from spent fuel under varying concentrations of nitric acid and of uranium as well as under simulated pressurized heavy water reactor spent fuel feed conditions. At 50 g/L U in 4 M HNO3, DPu values for 1.1 M DHOA and 1.1 M TBP solutions in n-dodecane were 7.9 and 3.8, respectively. In contrast, significantly lower DPu value at 0.5 M HNO3 (4 × 10−3) for DHOA as compared to TBP (4 × 10−2) suggested that it was a better choice for coprocessing of spent nuclear fuel. This behavior was attributed to the change in stoichiometry of extracted species at lower acidity vis-a-vis the higher acidity. These studies suggest that plutonium fraction can be enriched with respect to uranium contamination in the product stream. DHOA displays better extraction behavior of plutonium and stripping behavior of uranium under simulated feed conditions. DHOA appears distinctly better than TBP with respect to fission product/structural material decontamination of U/Pu.